Progress in Nuclear Energy最新文献_第6页

Application of similarity analysis method in neutronics design of multi-purpose experimental reactor

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2025.105604

Chengjian Jin , Shichang Liu , Mengsen Zhang , Xingbo Wang , Mancang Li , Xuesong Yan , Tiancheng Lai , Yixue Chen

In response to the demands of new types of reactors and digital reactor development, as well as software validation, the core neutronics design of multi-purpose experimental reactors with high flexibility and broad applicability are conducted. This aims to enhance the critical physics experimental efficiency of new types of nuclear energy systems. To ensure the high credibility of experimental results, it is imperative to establish a high degree of similarity between multi-purpose experimental reactor cores and target reactor cores. Introducing similarity evaluation indexes based on energy spectrum, sensitivity, and uncertainty, a similarity analysis code is developed and validated through comparison with the similarity analysis code TSUNAMI-IP in the SCALE code system, yielding relative deviations in similarity calculations all below 1%. Design schemes for core loading in multi-purpose experimental reactors are explored, proposing driver core and experimental core design scheme. The experimental core can be flexibly adjusted according to different target cores to accommodate various fuel assembly types, such as plate, annular, and rod configurations, as well as diverse experimental requirements for fast and thermal spectrum. Utilizing the developed similarity analysis code, the similarities between the experimental core and target cores in different loading schemes are analyzed. Furthermore, optimization designs for loading schemes are conducted to reduce the number of experimental core assemblies while satisfying similarity conditions, thereby enhancing core economics. The realization of the design goal of one reactor for multiple purposes with flexible adjustments reduces the development cycle and costs of experimental reactors, thereby improving their utilization efficiency. This research holds significant implications for the development of multi-purpose experimental reactors and new types of nuclear energy systems.

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引用次数: 0

pyMAISE: A Python platform for automatic machine learning and accelerated development for nuclear power applications

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2024.105568

Patrick A. Myers , Nataly Panczyk , Shashank Chidige , Connor Craig , Jacob Cooper , Veda Joynt , Majdi I. Radaideh

Despite significant advancements in artificial intelligence and machine learning (AI/ML) algorithms and their potential in nuclear engineering applications, the field still lacks a framework that automates ML model development and deployment for nuclear engineering problems. To address this, pyMAISE (Michigan Artificial Intelligence Standard Environment) is introduced, which is a Python package that features automated hyperparameter tuning, model explainability, model training, validation, postprocessing, and deployment for various ML models relevant to nuclear engineering. pyMAISE provides a platform for researchers to demonstrate new models on benchmarked datasets and currently supports nine benchmark problems across reactor physics and design, reactor control, thermal hydraulics, fuel performance, safety analysis, and anomaly detection. In this work, pyMAISE is demonstrated in three applications: critical heat flux prediction, microreactor power prediction, and fault detection in electronic signals. pyMAISE provided efficient model search and performance results that meet or exceed other studies. For critical heat flux prediction, feedforward neural networks (FNN) and random forests were the top models achieving

R^{2} = 0.999

when six input features were used. FNN was the best performer for predicting microreactor quadrant power

R^{2} = 0.97

, 0.26 greater than the closest classical ML model. In fault detection, pyMAISE models achieved 81% test accuracy in detecting faulty signals using long short-term memory, which may prevent various accident scenarios that could cause facility downtime. As pyMAISE continues to develop through a multi-phase approach, the goal is to integrate uncertainty quantification and deployment tools to expedite the creation of explainable and licensable AI technologies for nuclear power plants.

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引用次数: 0

Multi-gene genetic programming of critical boron concentration and power peak factor for nuclear reactor fuel reload calculations

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2024.105596

Marcos A.G.S. Filho , Alan M.M. Lima , Victor H.C. Pinheiro

Nuclear fuel reload aims to search for a core configuration of partially burned and fresh nuclear fuel, optimizing the operational cycle length while assuring safety limits. For each configuration, operational cycle length and safety limits are evaluated in terms of boron concentration and power peak factor, respectively. Other safety parameters are not currently predicted with MGGP. In practice, a licensed numerical model is provided by the reactor manufacturer to estimate these physical parameters, and each configuration is simulated in approximately 300 s. Considering all possible core combinations, this approach becomes computationally unfeasible. This work introduces the Multi-Gene Genetic Programming (MGGP) to generate an explicit closed form mathematical function to estimate the nuclear reload physical parameters more efficiently. Results for a study of a single cycle of the Angra-I reactor show that the algorithmically generated mathematical function calculates boron concentration and power peak factor with a coefficient of determination of 0.997 and 0.95, respectively. For each configuration, the time of assessment is approximately 8E-4 s, which is several orders of magnitude faster than common licensed numerical tools, potentially enabling expensive optimization studies. Also, MGGP performance is compared with an implemented Artificial Neural Network (ANN), and model results are compared.

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引用次数: 0

Impact response analysis of fuel assembly based on nonlinear simplified dynamics modeling

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2025.105617

Tao Yang, Yixiong Zhang, Fengchun Cai, Ke Zhang, Zi Lu

As the lateral amplitude of the fuel assembly grows, it exhibits nonlinear characteristics, including a decrease in stiffness, a decrease in frequency, and an increase in the damping ratio. This study aims to simulate the aforementioned phenomena by introducing the Bouc-Wen nonlinear hysteresis model to develop a nonlinear mechanical model for the lateral vibration of fuel assemblies. Additionally, a multi-objective genetic optimization algorithm (NSGA-II) is employed to determine the model's unknown parameters. The computational results indicate that the nonlinear model effectively simulates the phenomena of decreased stiffness, decreased frequency, and increased damping ratio observed in the tests. Based on this nonlinear mechanical model, this study developed an impact analysis model for a single group of fuel assemblies using the Kelvin impact element. The model was then utilized to simulate fuel assembly single-point and multi-point impact tests. The calculation results indicate that the maximum impact force

F_{\max}

at each grid position closely aligns with the test results, exhibiting a maximum relative error of less than 2%. This proves that the impact analysis model effectively simulates the collision behavior of the fuel assembly. Simultaneously, the effects of parameters like impact stiffness, impact gap, and impact impulse on the computational results are also thoroughly researched in this study. Some conclusions that are helpful for fuel assembly impact simulation studies are summarized, such as the discovery that impact stiffness and impact gap have a significant non-linear effect on the grid's

F_{\max}

.

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引用次数: 0

Investigation on irradiated-thermal-mechanical coupling performance and failure behaviors of TRISO particle fuel for space reactor applications

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2025.105633

Lin Peng , Yuan Yang , Songyang Li , Zhonghao Nong , Simiao Tang

TRISO particles, recognized for their excellent ability to retain fission products and maintain structural integrity, are crucial candidates for space reactor applications. However, current studies primarily focus on conventional reactor conditions, leaving gaps in understanding their performance under the more demanding conditions of space reactors. This study aims to address these gaps by developing a thermal-mechanical coupling simulation for TRISO particles using the finite element method, incorporating comprehensive material properties, irradiation behaviors and fission gas release. The model was validated through tangential stress comparison in IAEA CRP-6 Case 8 and further verified with TRISO particle calculation results from the BISON code, demonstrating the reliability of the simulation and the accuracy of the model. Under space reactor conditions, the higher kernel power, prolonged fuel lifetime, and increased fission gas pressure significantly elevate the tensile stress in the CVD-SiC layer, raising its failure probability at the later stages of burnup. While the SiC layer remains highly reliable under conventional HTGR conditions, its failure probability under space reactor conditions may become substantial. Sensitivity analysis was conducted by varying five key parameters. The results revealed that higher temperatures, greater fission rates, and extended operation times significantly increase the buffer-IPyC gas pressure, posing the greatest challenge to TRISO particle performance in space reactors. These findings reveal the greatest challenges for TRISO particles in space reactors compared to conventional reactors and underscore the necessity of enhancing their designs to ensure reliability and structural integrity in space reactor applications.

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引用次数: 0

Neighboring country's nuclear power plant? “No, thanks”: An energy justice analysis

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2024.105559

Emrah Akyuz

The Metsamor nuclear power plant in Armenia, which was commissioned in 1976, is located 16 km from the Turkish border. The risks and opportunities associated with Metsamor, which meets 40% of Armenia's energy needs, also affect those living along the Turkish border; however, it is not academically known how these risks and opportunities are distributed among those living along the Turkish border. There are no comprehensive studies analyzing the impacts of the Metsamor nuclear power plant on Türkiye from an energy justice perspective. To address this gap in understanding, face-to-face interviews were conducted with residents (n = 44) living 30 km away from Metsamor on the Turkish border. Two main conclusions were reached: first, the participants perceive that the opportunities and risks of the Metsamor NPP are not shared fairly with those living in Turkey; and second, those living on the Turkish border near Metsamor do not have the right to participate in any decision-making processes related to nuclear energy, access to information, and access to justice. This study concludes that nuclear power plants located on a country's borders can lead to energy injustice in neighboring countries.

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引用次数: 0

Control of the VVER-1000 core power using optimized T-S fuzzy controller based on nonlinear point kinetic model

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2024.105560

Ahmed E. Salman , Magy M. Kandil , Afaf A.E. Ateya , Magdy R. Roman

It is important for nuclear power plants (NPPs) to have the capability to adjust their output to meet the requirements of the power system. This calls for the development of efficient control algorithms to preserve steady/safe operating conditions. The current research proposes an optimized Takagi-Sugeno (T-S) fuzzy control algorithm to regulate the core power level of a VVER-1000 reactor. In the research, a straightforward method of tuning and optimizing T-S fuzzy controller is presented. The designed controller has a nonlinear mathematical structure of PI controller which has parameters that interpolate smoothly among a number of linear classical PI controllers. T-S fuzzy controller gains are optimized using sequential quadratic programming (SQP). The objective function is defined as a weighted sum of a performance index, the integrated time absolute error (ITAE), and the stabilization time based on Lyapunov synthesis. Simulations are conducted based on a nonlinear point kinetic model of VVER-1000 reactor to evaluate the proposed approach. The performance is investigated in load-tracking mode at different rates and in the presence of external disturbances. The results demonstrate the superiority of the optimized T-S fuzzy control strategy in terms of tracking accuracy and responses to disturbances compared to conventional control methods.

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引用次数: 0

MC/MOC two-step method for reactor physics analysis of helical cruciform fuel reactor

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2025.105607

Qufei Song, Ruixiang Wang, Hui Guo, Hanyang Gu

Helical cruciform fuel (HCF) features superior thermal conductivity, high heat transfer area-to-volume ratio and continual mixing between channels, but its complex radial structure, axial structure and spectrum pose significant challenges to reactor physics analysis. This paper utilizes Computer-Aided Design based Monte Carlo method to analyze the impact of cruciform radial section and helical axial structure on reactivity, radial and axial power distribution. The accuracy loss of Constructive Solid Geometry layered helical modeling and non-helical modeling of HCF is investigated. Results indicate that the impact of helical structure on the k_eff is below 60 pcm, with the maximum effect on radial power of 0.7%, and on axial power distribution of 0.08%. Subsequently, a Monte Carlo/Method of Characteristics (MC/MOC) two-step method is developed and verified. Multi-group cross-section (MGXS) is generated by continuous-energy Monte Carlo (CEMC) calculation, corrected using super-homogenization (SPH) equivalence technique. 3D full HCF reactor core calculation is performed by 3D MOC. The results show that both the SPH corrected MGXS generated by CEMC and 3D MOC core calculation exhibit high accuracy. Compared to the CEMC results, the overall k_eff error of the MC/MOC two-step method is −109 pcm in 2D HCF core calculations and 225 pcm in 3D HCF core calculations.

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引用次数: 0

Consequence evaluation methodology for attacks on spent nuclear fuel dry casks

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2024.105591

Luke Lebel

Driven by lessons from the war in Ukraine, the “beyond design basis threat” may have to become as important concept to nuclear security as the “beyond design basis accident” has been post-Fukushima for nuclear safety. Military attacks on nuclear sites can lead to major radiological release scenarios that would be otherwise impossible. The explosive dispersal of spent fuel in dry cask storage is an example of this, and a methodology for assessing the consequences and protective action strategies are presented. Experience from past radiological dispersal device experiments are used to define the explosive aerosolization and subsequent atmospheric transport of the spent fuel particles. Amalgamated dose conversion factors are developed, based on the radionuclide inventory in aged spent fuel, that can be applied to any dispersion calculation with a unitary release. The high actinide content in the releases drive significant inhalation dose hazards during both the initial plume dispersal, as well as in the contaminated areas afterwards driven by particle resuspension. A set of derived airborne actinide concentration (

D A C

) limits are developed, to complement existing operational intervention levels for recommending evacuation or relocation of members of the public out of a contaminated area.

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引用次数: 0

Numerical research on the influence of inlet shape on the air flow and heat transfer characteristics inside the reactor channel

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2025-02-01 DOI: 10.1016/j.pnucene.2025.105620

Fulong Zhao , Xiaojian Ge , Zhifang Qiu , Ersheng You , Ruibo Lu , Sichao Tan , Ruifeng Tian

To improve the thermal efficiency of a reactor, changing the inlet shape of the core channel can be an effective approach. A deeper understanding of how inlet shape impacts local heat transfer characteristics is therefore essential. In this research, numerical simulations are employed to analyze the distribution characteristics of local heat transfer coefficients under different conditions and the influence of different inlet shapes on the local velocity field, temperature field and average heat transfer coefficient under a certain condition, and the influence mechanism of inlet shape on local heat transfer is summarized. The results show that inlet shape affects local heat transfer characteristics within a range of approximately 40–50 times the channel diameter, with this influence diminishing as inlet length increases; The overall average heat transfer coefficient of the heating section increases by 0.88%–3.3%, and the local average heat transfer coefficient at the inlet increases by 2.9%–5.6% after the adiabatic inlet is added before the heating section. Among various inlet geometries, the round box-shaped inlet demonstrates the least impact on the heat transfer coefficient in heat pipes when inlet lengths are same. Consequently, a round box-shaped inlet structure is recommended for the heating segment's inlet. This research provides a valuable reference for the local heat transfer characteristics of reactor core channels using compressible gas as the medium, as well as for improving the thermal efficiency of the core.

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引用次数: 0